The long and successful history of PUREX begs the question of its utility in the future of nuclear energy. In the current climate of developing a truly “closed,” “next generation,” or “advanced” nuclear fuel cycle, it is highly probable the first incremental improvements would be initially based on the capabilities of the current industrial standard, i. e. PUREX. Hot topics regarding the concepts of future fuel cycles are multi-faceted, but could be broadly categorized as containing two objectives: proliferation resistance (i. e., “no pure Pu” separation) and environmental concerns relating to radiotoxicity and disposal of the nuclear waste. The important capabilities of PUREX are the separation of major actinides prevalent in used nuclear fuel, primarily U, Pu, and Np (as well as the long lived and environmentally mobile fission product Tc). The next generations of PUREX processing will likely involve modification in the process chemistry to recover Np for recycle with the U and Pu into future mixed oxide (MOX) fuels and pre­clude the separation of a pure Pu stream by keeping some fraction of the U and/or Np with the Pu product. Another potential use of PUREX would be the separation of the minor actinides (MA), notably Am, by making use of the accessibility of higher oxidation states, albeit this use of PUREX is conceptual and much further down the road and consequently will not be further discussed here.

a 5 = superior; 4 = good; 3 = average; 2 = below average; 1 = poor. b Process flexibility includes such factors as the range of O/A flow ratio, the turndown in flowrate, and the ease with which the location of feed and product streams can be changed.

c Considered an advantage when process chemistry and kinetics requires long residence time.

d Considered an advantage when solvent degradation is a concern.

The process details discussed above indicate that there are several reasons to improve the current PUREX process. The distribution of neptunium between product streams that require additional decontamination, espe­cially the uranium product, is one of the major concerns. Currently, Pu is separated in high purity, converted to an oxide and subsequently back blended with U to prepare MOX. Intuitively, separating a combined U and Pu product (preferably in the appropriate and controllable ratios for MOX fuels) would substantially simplify the overall process, decrease the number of unit operations, and alleviate proliferation concerns. Also, there is inter-

est in introducing centrifugal contactors in advanced separation cycles due to the benefits of short residence time, small footprint, lower holdup volumes, and decreasing the risks from criticality issues. This technology obviously limits the chemical reactions to those with fast kinetics putting further constraints on the future use of PUREX processes.

Much work has been carried out in the United Kingdom and in France to improve PUREX into an “Advanced PUREX” or COEXTM process, respectively. This work has aimed at bleeding U (and possibly Np) into the Pu product to eliminate the pure Pu stream and to reduce the number of solvent extraction cycles. Furthermore, by controlling the neptunium oxida­tion state and providing a single route for the neptunium, the uranium purification cycle could conceivably be eliminated, simplifying the process and effectively decreasing the amount of waste (Taylor 1997). Using similar means to achieve an improved PUREX process, the path considered in the US is slightly different due to other political goals (Laidler 2001). The pos­sibility to dispose of uranium as low-level waste (Vandegrift 2004) without the need for uranium purification cycle(s) would require thorough decon­tamination of the uranium from plutonium and neptunium. This route has been dubbed the “UREX” process since the goal is URanium EXtraction and the process is very similar to the PUREX process but eliminates plu­tonium extraction such that the Pu would follow the remaining transuranic actinides.

162 Advanced separation techniques for nuclear fuel reprocessing